ML20086R484

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Amend 46 to License NPF-57,removing Tech Spec Table 4.4.6.1.3-1, Reactor Vessel Matl Surveillance Program - Withdrawal Schedule
ML20086R484
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 12/16/1991
From: Chris Miller
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20086R486 List:
References
NUDOCS 9201020080
Download: ML20086R484 (13)
v • d • e


Text

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'o UNITED ST ATES l 'y 3 #,

NUCLE AR REGULATORY COMMISSION n

wAmmates o c. rosa s, s.. f r ueL I C_g RyJ C E,,E L E C,1 RJ,C, 8, G AS, C 0ff,Afq S

AT ul; TIC CITY ELECTP.1C C014AfiY 00CLET t;0. EO-3E4 HOPECITEKGENERAQt;GSTAT10N WEHDf> Eft % M *[J,TL Of ERA,TI!i,G,3J chi,$E Amendment flo. 46 License flo, flPF-57 1.

The f uclear Regulatory Commission (the Cormission or the NRC) has found that:

A.

The application for amenament filed by the Public Service Electric &

Gas Company (PSE&G) dated October 10, 1991 complies with the standards and requirements of the Atomic Energ., Act of 1954, as an> ended / he t.ct), and the Cornission's rules and regulations set forth i'.e CFR Chapter 1; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the r.ommission; C.

There is reasonable assurance: (i) that the activities authorized by this aundment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter 1; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and s6fety of the public; and L.

The istsance of this aniendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.0.(2) of f acility Operating License tio. flPF-57 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment tio. 46, and the Environmental Protection Plan contained in l

Appendix B, are hereby incorporated in the license.

PSE&G shall operat0 the facility in accordance with the Technical Specifications and the j

Environmental Protection Plan, l

1 9201020000 911216 DR ADOCK 05000304 PDR

_ _. _ _ _ _ _ ~ _. _. _ _ - _ _ _. -. - _ _. _ _. _ _ _ _ _ _.. _. _ _ _ _ - _ _... _ _ _ - _ _ _.. _ _ _. _ _ _

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2 3.

This license amendinent is effective 80 of its date of issuance and shall be implernented within 00 days of the date of issuance.

TOR THE f,'UCLf AR REGULATORY C0f'HIS$10N (hcYtr.$

Charles L. Miller, Director Project Directorate 1-2 Division of Reactor Projects - 1/11 Office of fluclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

December le 1991 l

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e ATTACHMLNT.T.O..L.i.t.f _H.S.C..A.f!.E.N.D_M.E.N.T.NO 46.._

fecjijIy,pptgeJjyp,!jr!Nstpp,_utt:sz DOCKETti0,$p354 Replace the following pages of the Appendix "A" Technical Specifications with i

the attached pages. The revised pages are identified by Anendment number and contain vertical lines indicating the area of change. Overleaf pages provided to maintain document completeness.*

Remove insert xi xi i

xii xii*

3/4 4 21 3/4 4-21*

3/4 4-22 3/4 4-22 3/4'4-23 3/4 4-23*

3/4 4-24 3/4 4-24 B 3/4 4 5 B 3/4 4-5*

B 3/4 4 6 B 3/4 4-6 B 3/4 4 7 B 3/4 4-7*

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B 3/4 4-8 B 3/4 4-8 I

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTI0l{

PAGE 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System....................................

3/4 4-21 figure 3.4.6.1-1 Minimum Reactor Pressure Vessel Metal Temperature Versus Reactor Vessel Pressure...............................

3/4 4-23 Table 4.4.6.1.3-1 (0eleted).............................

3/4 4-24 Reactor Steam 0 cme........................................

3/4 4-25 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES..........................

3/4 4-26 3/4.4.8 STRUCTURAL INTEGR1TY......................................

3/4 4-27 3/4.4.9

.RESIOUAL HEAT REMOVAL Hot Shutdown...................................

3/4 4-28 Cold Shutdown.............................................

3/4 4-29 3/4.5 EMERGENCY CORE COOLING SYSTEMS i

3/4.5.1 ECCS - 0PERATING..........................................

3/4 5-1 3/4.5.2 ECCS -

SHUTD0WN...........................................

3/4 5-6 3/4.5.3 SUPPRESSION CHAMBER.......................................

3/4 5-8 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity.............................

3/4 6-1 Primacy Containment Leakage...............................

3/4 6-2 Primary Containment Air Locks.............................

3/4 6-5 L

HSIV Sealing System.......................................

3/4 6-7 L

Primary Containment Structural Integrity..................

3/4 6-8 L

Drywell and Suppression Chamber Internal Pressure.........

3/4 6-9 t

l l

HOPE CREEK xi Amendment No. 46 l

..i I

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Drywell Average Air Temperature...........................

3/4 6 10 Drywell and Suppression Chamber Purge System..............

3/4 6-11 3/4.6.2 DEPRESSURIZATION SYSTEMS Suppression Chamber.......................................

3/4 6-12 Suppression Pool Spray....................................

3/4 6-15 Suppression Pool Cooling..................................

3/4 6-16 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES......................

3/4 6-17 Table 3.6.3-1 Primary Containment Isolation Valves......

3/4 6-19 3/4.6.4 VACUUM RELIEF Suppression Chamber - Drywell Vacuum Breakers.............

3/4 6 43 Reactor Building - Suppression Chamber Vacuum Breakers................................................

3/4 6-45 3/4.6.5 SECONDARY CONTAINMENT Secondary Containment Integrity...........................

3/4 6-47 Secondary Containment Automatic Isolation Dampers.........

3/4 6-49 Table 3.6.5.2-1 Secondary Containment Ventilation System Automatic Isolation Dampers Isolatiun Group No.

19..................

3/4 6-50

-Filtration, Recirculation and Ventilation System..........

3/4 6-51 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Containment Hydrogen Recombiner Systems...................

3/4 6-54 Drywell and Suppression Chamber Oxygen Concentration......

3/4 6-55 3/4.7 PLANT SYSTEMS 3/4 ? 1 SERVICE WATER, SYSTEMS Safety Auxiliaries Cooling System.........................

3/4 7-1 Station Service Water System..............................

3/4 7-3 Ultimate Heat Sink........................................

3/4 7-5 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM..................

3/4 7-6

-I HOPE CREEK xii

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REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curves A and A' for hydrostatic or leak testing; (2) curves 8 and B' for heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curves C and C' for operations with a critical core other than low power PHYSICS TESTS, with:

a.

A maximum heatup of 100'F in any ono hour period, b.

A maximum cooldown of 100'F in any one hour period, c.

A maximum temperature change of less than or equal to 20'F in any one hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves, and d.

The reactor vessel flange and head flange metal temperature shall be maintained greater than or equal to 79'F when reactor vessel head bolting studs are under tension.

APPLICABILITY:

At all times.

ACTION:

With #:iy of the above limits exceeded, restore the temperature and/ar pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the ef fects of the out-of-limit conditic. on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SURVEILLANCE REQUIREMENTS 4.4.6.1.1

  • During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6.1-1 curves A and A',

B and B', or C and C' as applicable, at least once per 30 minutes.

t i

HOPE CREEK 3/4 4-21

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i REACTOR COOLANT SYSTEM SURVEILLANCE REQ'J1REMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curves C and C' within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality and at least once per 30 minutes during system heatop.

4.4.6.1.3 The reactor vessel material surveillance specimens shall be removed and examined, to determine changes in reactor pressure vessel material properties, as required by 10 CFR 50, Appendix H.

The results of these examinations shall be used to update the curves of Figure 3.4.6.1-1 based on the greater of the following criteria:

a.

The actual shift in reference temperature for plate material from heat SK3238-1 and weld metal 510-01205 as determined by Charpy impact test, or b.

The predicted shift in reference temperatures for plate material from heat SK3025-1 as determined by Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials."

4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 70 F:

a.

In OPERATIONAL CONDITION 4 when reactor coolaht sy; tem temperature is:

1.

1 100 F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

1 80'F, at least once per 30 minutes, b.

Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

HOPE CREEK 3/4 4-22 Amendment No. 46

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MINIMUM REACTOR PRESSURE VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE Figure 3.4.6.1-1 HOPE CREEK 3/4 4-23

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i HOPE CREEK 3/4 4-24 Amendment No. 46

.o 9

REACTOR COOLANT SYSTEM BASES 3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the offects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and stantup and shutdown operations.

The various cate used for design purposes are provided in Section (4.9)gories of load cycles of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so thst the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G ar.d ASME Code Section III Appen-dix G.

The curves are based on the RT and stress intensity f actor information NDT for the reactor vessel components.

Fracture toughness limits and the basis for compliance are more fully discussed in FSAR Chapter 5, Paragraph 5.3.1.5, " Frac-ture Toughness."

The reactor vessel materials have been tested to detennine their initial RT The rtnults of these tests are shown in Table B 3/4.4.6-1.

Reactor NDT.

operation and iesultant fast neutron. E greater than 1 MeV, irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, HDT.

based upon the fluence, phosphorus content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda-tiens of Regulatory Guide 1.99, Revision 1 " Effects of Residual Elements on Predicted Radiot'on Damage to Reactor Vessel Materials." The pressure / tempera-ture limit cutve, Figure 3.4.6.1-1, curves A', B' and C', includes an assumed shift in RT for the end of life fluence.

NOT The actual shift in RTHDT of the vessel matorial will be established period-ically during operation by removing and evaluating, irradiated flux wires installed near the inside wall of the reactor vessel in the core area.

Since theneutronspectraat'thefluxwiresandvesselinsideradiusareessentially[

identical, the irradiated flux wires can be used with confidence in predictin reactor vessel material transition temperature shift.

The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis of the flux wire date and recommendations of Regulatory Guide 1.99, Revision 1.

HOPE CREEK B 3/4 4-5

i.

REACTOR COOLANT SYSTEM 1

BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves C, and C'Ic testing have been provided to assure compliance with the trinimum and A and A', for reactor criticelity and for inservice leak and hydrostat temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

The number of reactor vessel irradiation surveillance capsules and the frequencies for removing and testing the specimens in these capsules are pro-vided in UFSAR Section 5.3 and Appendix SA.

,3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE.

The surveillance requirements are based on the operating history of this type valve.

The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1977 Edition and Addenda through Summer 1978.

The' inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR 3

Part 50.55a(g)(6)(i).

3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate tempera-ture indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternete method of coolant mixing be in operation.

HOPE CREEK B 3/4 4-6 Amendment No. 46

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1 BASES TABLE B 3/4.4.6-1 g'

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A REACTOR VESSEL TOUGHNESS h

HEAT / SLAB HIGHEST PREDICTED UNIRRADIATED MAX. EOL BELILINE WELD SEAM I.D.

OR UPPER SHELF RT a RT RT COMPONENT OR MAT

  • L TYPE HEAT / LOT CU(%)

PR MOT (*F)

MDT(*F)

(FT-LES)

ET(*F) t Plate SA-533 GR B CL.1 SK3025-1

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+19 20 76

+39 Weld Long. seams for D55040/

.08

.010

-30 17 135

-13 shells 4&S and girth 1125-D2000 i

weld between 4&S i

NOTE:

  • These values are given only for the benefit of calculating the end-of-life (EOL) RTMOT-i HEAT / SLAB HIGHEST REFERENCE NON-BELTLINE MT'L TYPE OR OR P ERATURE COMPONENT WELD SEAM I.O.

HEAT / LOT NOT (*F)

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Shell Ring Connected ta SA 533, GR.B. C1.1 All Heats

+19

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q Vessel Flange Botton Head Dome SA 533, GR.B. C1.1 All Heats

+30 1

t Bottom Head Torus SA 533, GR.B. C1.1 All Heats

+30 i

w LPCI Nozzles SA 508, C1.2 All Heats

-20 Top Head Torus SA 533, GR.B. C1.1 All Heats

+19 Top Head Flange SA 508 C1.2 All Heats

+10 Vessel Flange SA 508, C1.2 All Heats

+10 Feedwater Nozzle SA 508 C1.2 All Heats

-20 i

Weld Metal All RPV Welds All Heats 0

Closure Studs SA 540, GR.B. 24 All Heats Meet 45 f t-lbs & 25 mils i

lateral expansion at +10*F The design of the Hope Creek vessel results in these nozzles experiencing a predicted EOL ffuence at I

i 1/4T of the vessel thickness of 1.6 x 1017 n/cm2 Therefore, these nozzles are predicted ta have an i

EOL RT of -6*F.

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AT 1/4 T A5 A FUNCTION OF $ERVICE LIFE' l

Bases Figure B 3/4.4.6-1 l

  • At 90% of RATED THERMAL POWER and 90% availability HOPE CREEK B 3/4 4-8 Amendment No. 46

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